1,721,106 research outputs found
HERA (Heterogeneous Epithermal Resonance Absorption): un modello per il calcolo delle pin cells.
No full textIl sistema SCALE (RSICC, 2006), preso a riferimento come schema generale per il presente studio, è stato sviluppato, a partire dalla metà degli anni ’60, per la Nuclear Regulatory Commission presso l’Oak Ridge National Laboratory, riguardo alla necessità di avere una metodologia standardizzata per la progettazione dei contenitori di sicurezza del combustibile irraggiato. Pur essendo nato come codice per il calcolo di schermature, con il tempo è stato ampliato e corredato di moduli funzionali capaci di eseguire anche analisi di criticità. SCALE è dunque un sistema modulare in cui determinati moduli, detti funzionali, vengono richiamati nella sequenza riportata nell’input ed eseguono il calcolo richiesto.
Uno di questi moduli è appunto il NITAWL-II. Le principali funzioni di questo modulo sono due:
- calcolo delle risonanze attraverso il metodo dell’integrale di Nordheim;
- conversione di una libreria master, in formato AMPX, in una libreria di lavoro (working), ancora nello stesso formato, ma capace di essere letta da eventuali moduli successivi.
Per come è stato implementata, la trattazione delle risonanze mediante il metodo di Nordheim, consente la risoluzione dell’equazione del rallentamento in una regione contenente il nuclide assorbitore più due eventuali moderatori interni. Tale regione ancora una volta deve essere monodimensionale ovvero in geometria sferica, cilindrica o a slab, circondata da un mezzo moderante in cui si assume che il flusso abbia il valore asintotico.
Attraverso un opportuno fattore di Dancoff si può tenere conto anche in questo caso dell’eventuale presenza di ulteriori regioni assorbenti. Tale fattore, tuttavia, non è calcolato direttamente dal codice bensì deve essere introdotto dall’utente nel set di informazioni che descrivono il calcolo di risonanza stesso.
L’allargamento Doppler delle sezioni d’urto di cattura, fissione e scattering è effettuato tramite l’impiego di una opportuna tabulazione delle funzioni Psi e Csi di Doppler.
L’applicazione di questo metodo coinvolge tuttavia due approssimazione non indifferenti:
- ciascun nuclide, per il quale debba essere effettuato il calcolo di risonanza, viene trattato senza considerare in alcun modo le risonanze degli altri nuclidi coinvolti nello stesso calcolo, ovvero non è ammessa, in nessun caso, la possibilità di sovrapposizione delle risonanze tra i diversi nuclidi considerati, in pratica è come se ciascun nuclide non risentisse della depressione del flusso conseguente all’assorbimento delle risonanze degli ulteriori nuclidi che insieme ad esso compongono l’assorbitore;
- il flusso neutronico è considerato spazialmente uniforme sia nel moderatore che nell’assorbitore, si adotta dunque una risoluzione del tipo due regioni e due zone.
Si è sentita così l’esigenza, alla luce anche di alcuni problemi di calcolo mostrati dalla sequenza SCALE per reticoli contenenti percentuali non trascurabili di plutonio(D'Agata, 1997/1998), di realizzare un modulo, completamente indipendente ed originale, che consentisse la risoluzione dell’equazione del rallentamento nel campo delle risonanze epitermiche senza coinvolgere tutte le approssimazioni adesso osservate
Computational design and preliminary measurements of a mixed-field irradiation facility
No full textA boron neutron capture therapy (BNCT) facility was built at the R2-0 research reactor at Studsvik (Sweden). This facility was characterized by the presence of two different neutron filter/moderator assemblies: one optimized for clinical irradiation of patients affected by glioblastoma multiforme, a tumor of the central nervous system; the other conceived to provide a well-characterized mixed radiation field with variable ratios of high- and low-LET components. In this paper the design of the large irradiation cavity located at the end of the second beam will be described. The possibility to change the thickness of the heavy water moderator inside the filter makes it possible to vary the neutron spectrum inside the irradiation cavity to achieve different mixed-field conditions. The main goal of this work was to design the irradiation cavity to achieve a pure thermal neutron field when using the maximum thickness of D2O. Preliminary measurements performed inside the modified irradiation cavity confirmed the calculated results. The thermal neutron flux measured at the entrance of the irradiation cavity for a D2O thickness of 45 cm was equal to 2.63x10^9 cm-2 s-1 (2.0%, 1sd) in comparison with an estimated value of 2.71x10^9 cm-2 s-1 (5.1%, 1sd). In the case of 15 cm of D2O the measured thermal neutron flux was 5.18x10^9 cm-2 s-1 (2.4%, 1sd) in comparison with an estimated value of 5.25x10^9 cm-2 s-1 (5.0%, 1sd). According to the calculation the photon contamination inside the irradiation cavity is very low, however, no preliminary measurements were performed to confirm this data. Although both the nuclear research reactors located at Studsvik have been definitely shutdown, information on the design and on the results obtained are still relevant and could be useful to plan or compare similar installations elsewhere
A boundary element–response matrix method for the multigroup criticality problems in the SP3 approximation
No full textThe 3D xyz multigroup BERM-SP3 transport method is here presented. This method, based on the simplified spherical harmonics approximation (SPN), with N = 3, and the assumption of linearly anisotropic scattering, makes use of a Response Matrix (RM) solution procedure, coupled with the Boundary Element Method (BEM), by which the SP3 partial differential equations are reduced to a system of boundary integral equations in terms of partial currents. Numerical problems, all endowed with a complete set of data and the reference results obtained by well assessed transport codes such as the discrete ordinate code TORT and the Monte Carlo code MCNP, illustrate the accuracy and efficiency of the method
A Boundary Element - Response Matrix method for 3D neutron diffusion and transport problems
Full text linkAn application of a 3D Boundary Element Method (BEM) coupled with the Response Matrix (RM) technique to solve neutron diffusion and transport equations for multi-region domains is presented. The discussion is here limited to steady state problems, in which the neutrons have a wide energy spectrum, which leads to systems of several diffusion or transport equations. Moreover, the number of regions with different physical constants can be very large. The boundary integral equations concerning each region are solved via a polynomial moment expansion and the multi-fold integrals there involved are reduced to single or double integrals, taking advantage of suitable recurrence formulas. The usual unknowns (the boundary particle density and its normal derivative) are here replaced by the partial currents entering or leaving each cell. The intuitive physical meaning of such quantities facilitates the application of the response matrix technique. Only eigenvalue (criticality) problems will be here considered. As it regards the transport equation, the use of the so called Simplified Spherical Harmonics method allows, through suitable approximations, to cast the problem into a system of differential elliptic equations of the diffusion type, which can still be solved by BEM
SU-E-T-521: Investigation of the Uncertainties Involved in Secondary Neutron/gamma Production in Geant4/MCNP6 Monte Carlo Codes for Proton Therapy Application
Full text linkA major concern in proton therapy is the production of secondary neutrons causing secondary cancers, especially in young adults and children. Most utilized Monte Carlo codes in proton therapy are Geant4 and MCNP. However, the default versions of Geant4 and MCNP6 do not have suitable cross sections or physical models to properly handle secondary particle production in proton energy ranges used for therapy. In this study, default versions of Geant4 and MCNP6 were modified to better handle production of secondaries by adding the TENDL-2012 cross-section library
Radioterapie oncologiche alternative: la BNCT (terapia dei tumori per cattura neutronica del Boro)
No full textLa Terapia per Cattura Neutronica del Boro (BNCT: Boron Neutron Capture Therapy) è una terapia binaria per la cura dei tumori nella quale una delle due componenti necessarie per avere l’effetto desiderato è costituita dalla radiazione neutronica
Lumped Parameter Analysis of Autogenous Propellant Pressurization System for Nuclear Thermal Rocket
No full textThis work proposes a new propellant management configuration for an ammonia-fueled nuclear thermal propulsion system. The suggested configuration maximizes the advantage deriving from the autogenous pressurization of ammonia by exploiting the thermal power lost by the nuclear reactor toward the vacuum space due to leaking radiation. In this layout, a tank containing ammonia in saturated conditions is placed near the nuclear reactor and receives an input thermal power proportional to the dose of gamma rays and neutrons absorbed by the tank walls and the ammonia itself. Such a thermal power accelerates the vaporization process of the saturated ammonia, thus increasing the pressure in the tank. A pressure regulator valve exploits this overpressure to pressurize the ammonia propellant contained in a run tank to the level required by the mission by connecting the two ammonia volumes. The pressure achieved inside the run tank pushes the propellant with an adequate mass flow rate inside the nuclear reactor. The developed lumped parameter analysis shows how this propellant management system can provide a constant mass flow to the nuclear reactor without using a turbopump assembly. Moreover, it is shown how the proposed concept allows for a reduction in the radiation shield mass
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